The present invention relates to low temperature, lead-free, phosphate glass compositions which can be used to encapsulate and immobilize radioactive, hazardous and mixed wastes. Specifically, the present invention concerns the development of novel phosphate glass compositions having reduced processing temperatures and improved chemical durability when compared with existing conventional borosilicate and phosphate glass formulations. The new materials of the present invention have been developed primarily for use in vitrifying radioactive, hazardous and mixed wastes, but can also be used in other industrial glass applications.
Industrial, utility, military nuclear operations produce radioactive hazardous and mixed wastes which must be treated in an appropriate manner before disposal to prevent environmental contamination.
Vitrification has been selected for immobilization of high-level radioactive waste (HLW) because glass is highly stable, very durable and has the ability to incorporate a wide variety of chemical contaminants. However, requirements for storage and disposal of low-level radioactive (LLW) and mixed wastes differ from those for HLW. In addition, LLW and mixed waste contaminants have waste chemistries which can vary from those typically found in HLW. Thus, evaluation of vitrification or immobilization of these wastes must consider these differences.
In the past, most research on vitrification of mixed waste has focused on borosilicate glass formulations originally developed for HLW. Although borosilicate glass has demonstrated good long-term chemical durability and is both thermally and physically stable, it requires relatively high process temperatures (1200.degree.-1500.degree. C.) for effective encapsulation of waste. These high temperatures are a major drawback because of volatilization of certain isotopes such as .sup.99 Tc and .sup.137 Cs and heavy metals, such as Pb and Cd. To capture and stabilize the off-gas contaminants high temperature processes require the use of secondary treatment systems. Further, borosilicate glass processing is incompatible with even minor amounts (&gt;1.5 mole %) of P.sub.2 O.sub.5 in a waste stream. Insoluble phosphate phases form depending on the amount of CaO and rare earth oxides present. Further, wastes containing more than about 0.3 mole % Cr.sub.2 O.sub.3, more than about 16 mole % Al.sub.2 O.sub.3, or more than about 19 mole % Fe.sub.2 O.sub.3 have also been flagged as outside the borosilicate glass concentration envelope. Therefore, alternative vitrification processes or products are desirable for improved treatment of these wastes.
In recent years, most research on low temperature glass processes focused on phosphate glass. Phosphate glasses are network-forming materials, structurally similar to silicate glasses, whose basic building blocks are PO.sub.4 tetrahedra. Compared with silicate glasses, phosphate glasses offer significant advantages for waste vitrification, such as lower melting and softening temperatures and low melt viscosities. Use of phosphate glasses as final waste forms has been previously proposed, however interest in their development waned as durability of the glass matrix was found to be poor when compared with silicate glasses. Further, new process technologies and materials would have been required to handle their production.
Early work to develop phosphate glass as a nuclear waste form was pioneered at Brookhaven National Laboratory (BNL) under contract with the U.S. Atomic Energy Commission, beginning in the late 1950's and continuing through the late 1960's as described by Tuthill, et al. in "Development of the Phosphate Glass Process for Ultimate Disposal of High-Level Radioactive Wastes," BNL 50130(T-505) (1968). Work at BNL focused on development of a continuous glass-melting process for the treatment of Purex-type wastes, nitric acid solutions of the fission products and residual salts from the extraction of uranium and plutonium, together with corrosion products. Development of the BNL process was carried out through pilot-scale testing using simulated waste solutions. A plant-scale demonstration using actual Purex HLW was conducted during the mid to late 1960's as part of the Waste Solidification Engineering Prototypes (WSEP) program at the Pacific Northwest Laboratory (PNL) as described by McElroy, et al., in "Evaluation of WSEP High-Level Waste Solidification Processes" Waste Solidification Program Summary Report, Vol. 11, BNWL-1667 (1972).
In the BNL process aqueous Purex wastes were evaporated, forming primarily nitrate salts of Fe, Na, Al, Cr. and Ni, SiO.sub.2, SO.sub.4, and PO.sub.4 anions were also present in the waste, along with fission products, nominally in the concentration range of 10.sup.-3 M. Good glass formation was reportedly observed at P.sub.2 O.sub.5 compositions greater than 65 mole %, for the Fe.sub.2 O.sub.3 --Na.sub.2 O--P.sub.2 O.sub.5 ternary system, and at greater than 70 mol % for the Fe.sub.2 O.sub.3 --Al.sub.2 O.sub.3 --Na.sub.2 O--P.sub.2 O.sub.5 quaternary system. However, the Fe.sub.2 O.sub.3 --Na.sub.2 O--P.sub.2 O.sub.5 system, all glasses with less than 60 mole % P.sub.2 O.sub.5 devitrified. For the Fe.sub.2 O.sub.3 --Al.sub.2 O.sub.3 --Na.sub.2 O--P.sub.2 O.sub.5 system, all glasses with less than 70 mole % P.sub.2 O.sub.5 devitrified.
Glass formulations were melted at 1100.degree.-1200.degree. C., with temperatures reduced to 800.degree.-900.degree. C. during idling periods to prevent volatilization of phosphate. Metal ions were typically retained at more than 99.9% of their original amounts. Exceptions were ruthenium (94-99%) and cerium (99%), which appeared in the off-gas condensate, which was mainly H.sub.2 SO.sub.4. The melt vessel was constructed of Inconel 702 with a Pt liner. Corrosion of the Pt liner was not a problem, although cracks did develop on several occasions, corroding the container. Inconel in contact with the molten phosphate reacted to form elemental phosphorus which, in turn, combined with Pt to form alloys with lower melting points.
In Europe, in another example of a waste solidification process, phosphate glasses were investigated for solidification and metal embedding of liquid radioactive wastes as described by Van Geel, et al. in "Solidification of High Level Liquid Waste to Phosphate Glass-Metal Matrix Blocks," IAEA-SM-207/83, Vol. 1, pp. 341-359, (1976). Glass was melted at about. 1100.degree. C. using a refractory lined melter with a liquid waste feed. Glasses were prepared as small beads to avoid devitrification problems. Low-alkali or alkali-free iron-aluminum phosphate glasses were found to have improved chemical durability, even better than that of borosilicate glasses. Soxhlet (dynamic) leachabilities were found to be about 5.times.10.sup.-5 g/cm.sup.2 /d, although at temperatures above 100.degree. C., leachability increased rapidly. Glasses were nominally 5 wt % Al.sub.2 O.sub.3, 15 wt % Fe.sub.2 O.sub.3, and 50 wt % P.sub.2 O.sub.5, containing about 30 wt % waste oxides.
Lead-iron phosphate (LIP) glass nuclear waste forms were introduced and characterized in detail by Sales and Boatner in the mid 1980's. While lead phosphate glasses had been studied since the mid 1950's, their poor durability in water made them of little commercial interest. Vitrified waste forms containing iron oxide, however, were found to fare favorably compared with borosilicate glass waste forms, rekindling new interest in phosphate glass waste materials. LIP glasses could be processed at temperatures 100.degree. C.-250.degree. C. lower than borosilicate glasses using glass melting technology similar to that developed for borosilicate glasses. Solidified forms had dissolution rates in water about 1000 times lower than comparable borosilicate formulations, at 90.degree. C. in solutions with a pH between 5 and 9.
The addition of iron to lead phosphate glass was found to dramatically increase chemical durability of the glass, by a factor of about 10.sup.4 for a 9 wt % iron oxide addition. Also, the tendency for the glass to crystallize on cooling or reheating was greatly suppressed. Without the iron modifier, lead metaphosphate glasses completely crystallized in air at 300.degree. C. within a few hours. In contrast, LIP glasses were heated for several hundred hours at 500.degree. C. without any signs of devitrification. A variety of metal oxide modifiers were investigated for similar effects MgO, Al.sub.2 O.sub.3, CaO, Sc.sub.2 O.sub.3, TiO.sub.2, VO.sub.2, Cr.sub.2 O.sub.3, MnO.sub.2, CoO, NiO, Cu.sub.2 O, ZnO, Ga.sub.2 O.sub.3, Y.sub.2 O.sub.3, ZrO.sub.2, In.sub.2 O.sub.3, La.sub.2 O.sub.3, CeO.sub.2, and Gd.sub.2 O.sub.3), although none were as effective. Compositional ranges for the three major components were: PbO (37-60 wt %), Fe.sub.2 O.sub.3 (6-13 wt %), and P.sub.2 O.sub.5 (32-44 wt %), where waste loadings ranged from 14-19 wt %. Practical concentration ranges for glass matrix formation were given to be: PbO (40-66 wt %), Fe.sub.2 O.sub.3 (0-12 wt %), and P.sub.2 O.sub.5 (30-60 wt %). Glasses were melted at temperatures between 900.degree. and 1050.degree. C.
Structure of LIP glass was characterized extensively, using Mossbauer spectroscopy, electron paramagnetic resonance, Raman and infrared spectroscopy, EXAFS (Extended X-ray Absorption Fine Structure), low-angle X-ray scattering, liquid chromatography, and transmission electron microscopy. For glasses prepared below 900.degree. C., iron was incorporated in the glass as Fe.sup.+3. Average polyphosphate chain length was reduced from more than 15 for iron-free glasses, to about 2.6 for glass with 9 wt % Fe.sub.2 O.sub.3. Thus the improvement in durability was related to strengthened cross bonding between polyphosphate chains.
PNL conducted an evaluation of LIP glasses as part of the Second Generation HLW Technology Subtask of the Nuclear Waste Treatment Program. Their review found LIP glasses to have substantially better chemical durability than borosilicate glass, although severe devitrification (leading to reduced chemical durability) would result if glass waste forms were prepared in large canisters. A processing method would be required to rapidly cool the material to quench the vitreous structure. Similarly, LIP glasses were examined for their potential applicability for Savannah River Plant (SRP) waste. Phosphate glasses were found to be highly durable, however, the glass melts were highly corrosive with existing glass melting equipment. Thermal stability and waste component solubility were lower than for high-silica glasses. As a result, borosilicate glasses were found to be, overall, more favorable for SRP waste treatment.
Ultra low melting temperature lead-tin fluorophosphate glasses were investigated by Tick, P. A. in "Water Durable Glass with Ultra Low Melting Temperatures," Physics and Chemistry of Glass, Vol. 26, No. 6, pp. 149-154 (December 1984) showing glass transition temperatures of 75.degree. C.-150.degree. C. with good resistance to water attack. Reported compositional ranges investigated were 50-61 atomic percent (atomic %) Sn, 3.0-5.7 atomic % Pb, and 34-48 atomic % P, with fluorine and oxygen contents of 35-74 and 114-149 atoms per 100 cations, respectively. Glass melting was done in air at 450.degree. C. Although lead additions had little effect on glass transition temperature, lead-tin phosphate glasses were readily devitrified without lead as glass component, suggesting that lead has a significant effect on the bonding character in the glass. With regard to chemical durability, four distinct corrosion behaviors were noted for the glass formulations tested, corresponding to very high (&gt;10 mg/cm.sup.2 day) to relatively low (&lt;0.1 mg/cm.sup.2 day) dissolution rates. Best durability was associated with the compositional window defined by 8&lt;Sn/Pb&lt;13 and 1&lt;Sn/P&lt;2, with F/Sn=1.+-.0.2. Outside this region, phase separation was suspected, producing swelling or accelerated dissolution. There are no reports about their chemical stability in acid or basic solutions so far.
Few recent advances in phosphate glass waste forms have been reported as current interest seems to be focused on phosphate glass laser development and associated optical and electronic characterization. Structure models continue to be updated, using high performance liquid chromatography and x-ray spectrometry to analyze glass samples. Chemical durability improvements in alkali phosphate glasses were proposed in a patent application by Day and Wilder in "Chemically Durable Phosphate Glasses and a Method for Their Preparation; Patent Application," Department of Energy, Washington, D.C., PAT-APPL-6-447 847, 1982. Glasses containing 10-60 mole % of Li.sub.2 O, NaO, or K.sub.2 O; 5-40 mole % of BaO or CaO; 0-10 mole % of Al.sub.2 O.sub.3 ; and 40-70 mole % of P.sub.2 O.sub.5 were improved by incorporating up to 23 wt % of nitrogen. Nitrides were the favored additives.
Other phosphate containing formulations proposed for nuclear waste immobilization have been glass ceramics. Certain glass-ceramics or glass-composites may possess higher chemical durabilities than single phase glasses with respect to dissolution in water or corrosion resistance in harsh environments. Glass-ceramics or glass-composites may be formulated for vitrification, consisting of a major, thermodynamically stable crystalline phase and a relatively durable, vitreous matrix.
Various combinations of phosphate glass-ceramic matrices have been manufactured at research level. For example, new castable glass-ceramic dental materials have the potential for use in many phases of restorative dentistry. U.S. Pat. Nos. 3,732,087 and 4,431,420 disclose work which been carried out at Corning Glass Works on glass-ceramics based on crystallization of the parent glass. Bio-compatible calcium phosphate glass-ceramics were developed in Japan, possessing extremely high toughness and flexibility. These new materials were described as being durable in both acid and alkaline solutions by Abe, Y., et al. in "Calcium Phosphate Glass Ceramics for Biomedical and Biotechnological Applications," Asahi Glass Foundation for Industrial Technology, PB91-167056, PC A16/MF A02, (1990). In Canada, glass-ceramics based on partial crystallization of precursor glasses from the system Na.sub.2 O--Al.sub.2 O.sub.3 --CaO--TiO.sub.2 --SiO.sub.2 --X (where X is waste oxides or minor processing additives) were developed for the possible solidification of fuel recycle waste was described by Hayward, P. J. in "Review of Progress in the Development of Sphere-based Glass Ceramics," Scientific Basis for Nuclear Waste Management, IX Vol 50, Ed L. O. Werme (1986). Temperatures&gt;1250.degree. C. were required to create the glass; the ceramic phase was then recrystallized during sustained heating at 900.degree.-1050.degree. C. Similar results were obtained in the U.S. for R.sub.2 O--RO--P.sub.2 O.sub.5 and R.sub.2 O--Al.sub.2 O.sub.3 --P.sub.2 O.sub.5 glasses (where alkali metals are Na or Li, and alkaline earth metals are Ca or Ba) doped with 1 mol % of TiO.sub.2, ZrO.sub.2, Y.sub.2 O.sub.3, La.sub.2 O.sub.3, or Ta.sub.2 O.sub.5. Glass-ceramics from the system SiO.sub.2 --Cs.sub.2 O--Al.sub.2 O.sub.3 --(La, Ce).sub.2 O.sub.3 --P.sub.2 O.sub.5 --Zr were described for high level waste immobilization at processing temperatures of &lt;1600.degree. C. in U.S. Pat. No. 4,314,909 to Beall, G. H., et al. While these glass-cerarnics exhibit good chemical durabilities, high melting temperatures and associated losses of volatile fission products from the melts would restrict potential commercial use of these materials.
The glass compositions and vitrification process described above have many drawbacks. Processes relying on borosilicate glass formulations are incompatible with certain single-shell tank waste components, such as phosphates, calcium oxide and rare earth oxides (Wiemers, et al, PNL-7918 Report, p. 4.5). Moreover, borosilicate compositions require high-temperatures for processing which causes high loss of volatile hazardous metals requiring additional expensive off gas treatment. While lead-iron phosphate glasses appear to offer advantages of reduced process temperature and improved chemical durability compared with borosilicate glass, their lead content is considerable (40-60 wt %), posing the potential for excessive toxic releases from the glass matrix.
Accordingly, there is still a need in the art of waste disposal for a composition and method for encapsulation and immobilization of radioactive, hazardous and mixed wastes.
It is, therefore, an object of the present invention to provide a composition and process for permanent storage radioactive, hazardous and mixed wastes by encapsulation in low temperature, lead-free phosphate glass compositions. Another object of the present invention is to provide durable waste forms which can withstand delocalization by ecological forces. Another object of the present invention is to develop an encapsulating process which does not cause volatilization of hazardous material.